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This article presents a new finite-difference scheme for convection-diffusion equations for numerical prediction of heat transfer and fluid flows. This scheme, called efficient finite differencing (EFD), is an extension of "locally exact" one-dimensional analytical methods. It takes into account more accurately the distribution of source term and diffusion coefficient within the control volume. EFD has been applied to the set of sample problems including one- and two-dimensional steady-state and transient convection-diffusion transport. A comparison with exact analytical solutions as well as the results obtained using other popular methods is also given. The EFD scheme has been shown to be more accurate while using a mesh system with fewer grid points. EFD also shows good stability for most of the problems considered. [ABSTRACT FROM AUTHOR]

A numerical study on modeling of cross flow pressure drop coefficient for wire-wrapped triangular rod array has been performed. The objective is to clarify and also to extend the current modeling of cross flow pressure drop coefficient employed in subchannel analysis method. This study includes both pure laminar and fully turbulent flow conditions. Firstly, simulations employing pin pitch to diameter ratio ( P / D ) of 1.21 are carried out, and the results are compared with available experimental data in terms of wire correction parameter. The results for P / D = 1.21 are in good agreement with experimental data. Calculations employing smaller values of P / D , representing tight lattice rod configuration, have also been carried out. Based on the numerical results, new correlations of cross flow pressure drop are proposed. The correlations can be used to evaluate cross flow pressure drop coefficient in subchannel analysis codes. [ABSTRACT FROM AUTHOR]

Abstract: Bent pipe wall thinning has been often found at the elbow of the drain line and the high-pressure secondary feed-water bent pipe in nuclear reactors. Liquid droplet impingement (LDI) erosion could be regarded as one of the major causes and is a significant issue of the thermal hydraulics and structural integrity in aging and life extension for nuclear power plant safety. In this paper a computational methodology is established for simulation of LDI erosion using computational fluid dynamics (CFD) simulation and theoretical calculation. Two-phase flow numerical simulations are conducted for standard elbow geometry, typically with the pipe diameter of 170mm. This computational fluid model is built up by incompressible Reynolds Averaged Navier–Stoke equations using standard k–ɛ turbulence model and the SIMPLE algorithm, and the numerical droplet model adopts the Lagrangian approach. The turbulence damping in vapor–droplets flow is theoretically analyzed by a damping function on the energy spectrum basis of single phase flow. Locally, a droplet impact angle function is employed to determine the overall erosion rate. Finally, the overall and local investigations are combined to purpose a general methodology of LDI erosion prediction procedure, which has been complemented into CFD code. Based on our more physical computational results, comparison with an available accident data was made to prove that our methodology could be an appropriate way to simulate and predict the bent pipe wall thinning phenomena. [Copyright &y& Elsevier]

Abstract: The correct evaluation of flows at transitional Reynolds number in nuclear reactors is gaining higher importance in relation to the accident analysis for buoyancy-driven flows which dominate the heat decay removal process. In the present paper a comparative study of different turbulence modeling and wall treatment for the evaluation of a fluid flow in transitional Reynolds number, is presented employing computational fluid dynamics (CFD). The relative performance of the models is assessed through benchmarking of fully developed pipe flow at Reynolds number 4900 and of a 90° bend pipe at Reynolds number 5000. Predictions of velocity profiles at different locations are compared to both experimental and accurate numerical simulations. It has been found that the predictions between the models can vary considerably in particular in relation to the different wall treatment employed on the wall. The results show the concerns about the employment of the available turbulence models and wall treatments in low Reynolds number flow regimes and explanation is provided in relation to their formulation. [Copyright &y& Elsevier]

Abstract: A new method is proposed to separate the sodium void reactivity of step type FBR cores to components including non-leakage terms and a leakage term by using a newly developed perturbation code MCPERT where fluxes and adjoint fluxes are derived from a group-wise Monte Carlo code. The step type FBR core is a core where the height of the inner core is smaller than that of the outer core and a large sodium plenum region is located above the core so as to decrease the sodium void reactivity. The conventional diffusion perturbation method cannot treat such a large void region due to the diffusion approximation, while the Monte Carlo code can treat it exactly. In this study, a group-wise Monte Carlo code GMVP with a 70-group constant set JFS-3-J3.3 is employed to evaluate the neutron fluxes and adjoint fluxes which are used as inputs to the MCPERT code to evaluate the non-leakage terms. The leakage term is derived from the difference of the total sodium void reactivity evaluated by the direct calculation of GMVP and the summation of the non-leakage terms. It is found that the proposed method can provide the result approximately consistent to the ratio of the reactivity components derived from the conventional method. [Copyright &y& Elsevier]

Abstract: The bent pipe wall thinning has been often found at the elbow of the drain line and the high-pressure secondary feed-water bent pipe in the nuclear reactors. The liquid droplet impingement (LDI) erosion could be regarded to be one of the major causes and is a significant issue of the thermal hydraulics and structural integrity in aging and life extension for nuclear power plants safety. In this paper two-phase numerical simulations are conducted for standard elbow geometry, typically the pipe diameter is 170mm. The turbulence attenuation in vapor-droplets flow is analysed by a damping function on the energy spectrum basis of single phase flow. Considering the vapor turbulent kinetic energy attenuation due to the involved droplets, a computational fluid dynamic (CFD) tool has been adopted by using two-way vapor-droplet coupled system. This computational fluid model is built up by incompressible Reynolds Averaged Navier–Stoke equations using standard k–ε model and the SIMPLE algorithm, and the numerical droplet model adopts the Lagrangian approach, a general LDI erosion prediction procedure for bent pipe geometry has been performed to supplement the CFD code. The liquid droplets diameter, velocity, volume concentration are evaluated for the effects of carrier turbulence attenuation. The result shows that carrier turbulence kinetic energy attenuation is proved to be an important effect for LDI erosion rate when investigating the bent pipe wall thinning phenomena. [Copyright &y& Elsevier]

Abstract: In subchannel analysis, the conservation equations are solved for each channel in a complex fuel bundle, where the effects of fluid exchange between each subchannel are considered. The fluid exchange is commonly referred to as that caused by cross flow. Void drift is considered to be phenomenon resulting from attaining a hydrodynamic equilibrium state. Its mechanism has not been clarified, and the transport due to void drift is therefore estimated through empirical models in conventional subchannel analyses. Therefore, mechanistic model for the void drift phenomenon is required to apply the subchannel analysis to a variety of fuel bundle geometry. In this study, multi-dimensional analysis using two-fluid model was applied to two-phase flow inside a geometry simulating fuel bundle subchannels, for the purpose of clarifying the void drift mechanism. The comparison between the results of the numerical analysis and the experiment confirmed that the reliability of the numerical method used in this study. In this paper, a mechanistic model based on the Stanton number, which expresses the void diffusion coefficient based on the Lahey''s proposal, was proposed. [Copyright &y& Elsevier]

This study is aimed at the development of a fuel cycle concept for host countries with a lack of nuclear infrastructure. Two interrelated criteria, proliferation resistance and high-burnup, form the general framework of the fuel management scenario with the highest priority given to light water reactor technology and plutonium-free fresh fuel. Logically it implies the use of uranium oxide with enrichment close to 20%, whose effective utilization forms the main subject of the present paper. A sequence of two irradiation cycles for the same fuel pins in two different light water reactors is the key feature of the advocated approach. It is found that the synergism of PWR and pressure tube graphite reactor offers fuel burnup up to 140 GWd/tHM. Being as large as 8% in the final isotopic vector, the fraction of 238Pu serves as an inherent protective measure against plutonium proliferation. [Copyright &y& Elsevier]

• Adjoint sensitivity method is developed for high-Pr fluid natural circulation system. • High accuracy and high efficiency of adjoint method are demonstrated. • Applicability for high-Pr fluid system is verified using existing facility. Various passive safety systems have been developed for advanced nuclear reactors. The use of natural driving force for passive systems requires accurate assessment of reliability and performance. Sensitivity analysis is an efficient technique to test the reliability and performance of a safety system, and in order to improve the computational efficiency, the adjoint-based sensitivity method has been introduced. The present study aims to develop an adjoint-based sensitivity method to evaluate the sensitivity of natural circulation system using high-Pr fluid, and to assess its applicability to the experimental facility developed in UNIST. The conservation equations for a closed rectangular loop are written, and the corresponding adjoint system is developed based on the Lagrangian approach. Three different functionals are evaluated which represent thermal-hydraulics and heat transfer performance with respect to various input parameters. The developed adjoint method shows reasonable accuracy in the sensitivity analysis compared with other methods, while significantly reducing the computational time. Following sensitivity analyses using various design parameters assess the sensitivity of natural circulation mass flux, temperature distribution, and Nusselt number. Three different system conditions are imposed to investigate the effects of (i) the implementation of temperature-dependent fluid properties, (ii) the orientations of the heat exchanger and (iii) the operating temperature range, on the entire system reliability. It is found that the variation of fluid properties with respect to the temperature has an important effect on the stability of heat transfer performance in the system. Finally, the advantages in using high-Pr fluids for natural circulation heat transport system is verified by comparing the sensitivities of Nusselt number for different fluids with different Prandtl numbers. The outputs of this study are expected to provide the assessment of the reliability and performance of natural circulation system using molten salt, which can be used for further design studies on molten salt system with low computational burden. Furthermore, this paper demonstrate the applicability of the adjoint-based method for molten salt system applications. [ABSTRACT FROM AUTHOR]

Summary: The strongly coupled behaviors between neutronics and thermal‐hydraulics of liquid‐fueled molten salt reactors make it difficult to evaluate system behaviors, due to the transport of precursors along moving fuel. Extending an adjoint‐based method on the multiphysics approach, different assumptions on temperature dependencies of nuclear and thermophysical properties of salt are included in the local sensitivity analysis of a circulating liquid fuel system. Local sensitivity of various types of system response in steady‐state is analyzed for 39 parameters including coupling models, reactor design values, and kinetic constants of delayed neutron and decay heat precursors for a simplified 1D model of molten salt fast reactor. Extended adjoint‐based sensitivity analysis method for MSR is successfully validated achieving 1.38% deviation on average between a recalculation and adjoint method, comparing local sensitivities to all parameters. Also, it takes 66.3 times less in computational time compared with the recalculation method for evaluating the sensitivity of the same type of system response. The importance of all the parameters to the system response is analyzed according to the assumptions on temperature dependencies to nuclear data and salt properties. The most influencing ones are fission energy‐related terms, and their importance increases when temperature dependencies are taken into account, compared with constant properties. Changes of influences on the sensitivity are investigated from the relative changes of the parameter values in various system response types, and it implies the importance to consider the multiphysics modeling on the local sensitivity analysis. [ABSTRACT FROM AUTHOR]

Abstract: An inherently safe core concept with metallic fuel for sodium cooled fast reactor is proposed that has a negative void reactivity at the loss of coolant events without scram as well as a small excess reactivity during the operation cycle. The relationship of sodium void reactivities and burn-up reactivities to different core configurations has been studied quantitatively to clarify the core concept for large metallic fuel reactors. It has shown that the sodium void reactivity is greatly dependent on the core shapes while the excess reactivity is on the fuel compositions. It has also indicated that the core configuration that enables to enhance the neutron streaming through the region above the active core at coolant voiding is most effective to decrease sodium void reactivity. A 3000 MWt core composed of the flat inner core and annular outer core where the fuel volume fraction is relatively high and the sodium plenum is placed just above the active core region has been selected as a candidate core. The maximum excess reactivity of the candidate core at UTOP is about 0.4 $ and it can be reduced to approximately zero by power or inlet temperature adjustment during the operation cycle, meanwhile the sodium void reactivity is as low as -1.3 $ in negative that is enough to prevent ULOF sequences. [Copyright &y& Elsevier]

Abstract: As a part of the advanced subchannel code development project sponsored by Ministry of Economy, Trade and Industry, Japan, this paper describes improvement of the equilibrium void distribution model that is a main part of the vapor–liquid cross flow model. The three-component cross flow (TCCF) model is defined as the present framework that separates contributions of diversion, turbulent mixing and void drift. The Lahey''s void settling model is introduced to express the latter two components. Based on the high-resolution air–water database and other published data of steam-water tests, general trends of vapor–liquid cross flow processes are examined. It can be assumed that subchannel void distributions are dominated by the three major effects, i.e. the fluid dynamic effect, the geometrical effect and the narrow gap effect. The equilibrium void distribution model is modified to include the above-mentioned three effects. Three characteristic parameters are assigned for each of the three effects and they are identified experimentally as functions of the void fraction. Multi-dimensional lattice geometries are incorporated based on the two-dimensional flow network model. The network equation is constructed by mapping the equilibrium void balance problem into the force-deflection problem. The resultant models are verified based on equilibrium void distribution data obtained by Sadatomi and Kawahara. [Copyright &y& Elsevier]